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Category:OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING
MONTHYEARML20211D1321999-08-20020 August 1999 Application for Amend to License DPR-50 to Revise TS Bases Re Degraded Undervoltage Relay Setpoint Calibration Frequency & Degraded Voltage Relay Tolerance Rev ML20210J0941999-07-29029 July 1999 Suppl TS Change Request 274 to License DPR-50,revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance Tss,As Result of NRC Generic Concerns Re CR Habitability Issues ML20196K5901999-06-29029 June 1999 LAR 285 for License DPR-50,clarifying Authority in Encl License Pages to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-1 License to Amergen,Radioactive Matls Can Continue to Be Moved ML20209C0551999-06-29029 June 1999 LAR 77 for License DPR-73,clarifying Authority in Encl License Pages to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-1 License to Amergen,Radioactive Matl May Continue to Be Moved ML20195E3341999-06-0404 June 1999 TS Change Request 265 to License DPR-50,modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20209B5891999-05-26026 May 1999 LAR 281 to License DPR-50,revising TMI-1 UFSAR Chapters 5 & 14 to Permit Use of Conservative Deterministic Failure Margin Methodology for Seismic Analysis of Portions of Auxiliary Steam Line.Rept 240046-R-001 Encl.Rept Withheld ML20206R1111999-05-13013 May 1999 Application for Amend to License DPR-50,requesting That License Sections 2.a.(3) & 2.c.(7) & TS Section 3.1.1 Be Revised to Incorporate Administrative Updating & Changes to Bases Statement ML20205H0691999-04-0101 April 1999 TS Change Request 262 to License DPR-50,revising TS by Adding LCO Action Statements & Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20203B0421999-02-0202 February 1999 TS Change Request 274 to License DPR-50,expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems ML20196G4401998-12-0303 December 1998 LAR 278 to License DPR-50,requesting Consent to Transfer & Authorize Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License.Supporting Documentation,Encl ML20196H9751998-12-0303 December 1998 LAR 279 to License DPR-50,reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196G7671998-11-25025 November 1998 TS Change Request 277 to License DPR-50,changing Surveillance Specs for OTSG ISIs for TMI-1 Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only ML20154Q6201998-10-19019 October 1998 TS Change Request 248 to License DPR-50,adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18, Remote Shutdown Sys ML20154P8441998-10-19019 October 1998 Application for Amend to License DPR-50,providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage.Proprietary & non- Proprietary Calculation,Encl.Proprietary Info Withheld ML20154M3691998-10-15015 October 1998 Application for Amend to License Amend Request 276 to License DPR-50 Revising UFSAR Chapter 14 Accident Analysis Atmospheric Dispersion Factors.Ufsar Revised Pages Encl ML20249B2391998-06-11011 June 1998 TS Change Request 273 to License DPR-50,incorporating Alternate High Radiation Area Control Consistent W/Reg Guide 8.38 ML20217E5131998-03-23023 March 1998 Application for Amend to License DPR-50,revising TS Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 Efpy.Request Also Submitted in Response to NRC ML20217G0161997-10-0303 October 1997 Suppl to License Amend Request 268 to License DPR-50, Replacing Pages 4-80 & 4-81 Previously Submitted w/TMI-1 License Amend Request 268 on 970812 ML20211C3211997-09-19019 September 1997 Supplemental Application for Amend to License DPR-50 Re License Amend Request 269 Submitted on 970814.Suppl Includes Addl Change to TS Section 3.1.4 & Proposed Page Rev Clarifying That UFSAR Analysis Is More Conservative ML20211C2221997-09-19019 September 1997 Supplemental Application for Amend to License DPR-50 Re TS Change Request 266.Supplement Provides for Phased Review & Issuance of License Amends & Involves Administrative Correction to TS Decay Heat Removal Sys Leakage Rate ML20210J9951997-08-14014 August 1997 Revised License Amend Request 269 for License DPR-50, Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J3981997-08-12012 August 1997 Tech Spec Change Request 268 for License DPR-50,revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operations ML20198E7871997-07-30030 July 1997 TS Change Request 266 to License DPR-50,incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 ML20141E1311997-05-0808 May 1997 Application for Amend to License DPR-50,consisting of Change Request 264,incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2351997-04-21021 April 1997 Application for Amend to License DPR-50,requesting Rev to TS 3.3 Re Emergency Core Cooling,Reactor Bldg Emergency Cooling & Reactor Bldg Spray Systems ML20134M1111997-02-0707 February 1997 Application for Amend to License DPR-50,incorporating Certain Improvements from Revised STS for B&W Plants, NUREG-1430 & Addl Changes as Described in Encl TS ML20133D2651996-12-24024 December 1996 Application for Amend to License DPR-50,requesting Reinstatement of Auxiliary & Fuel Handling Bldg Ventilation Sys TS ML20132F3221996-12-16016 December 1996 Application for Amend to License DPR-50,consisting of Change Request 260,reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135C5201996-12-0202 December 1996 TS Change Request (Tscr) 75 to License DPR-73,relocating Audit Frequency Requirements to Pdms QA Plan ML20117H0181996-08-29029 August 1996 Application for Amend to License DPR-50,consisting of TS Change Request 257,incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1901996-06-28028 June 1996 Application for Amend to License DPR-50,consisting of Change Request 259,allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8751996-05-24024 May 1996 Errata to Application for Amend to License DPR-50,consisting of TS Change Request 243,replacing Pages for App a ML20101R1361996-04-10010 April 1996 Application for Amend to License DPR-50,consisting of Change Request 243,revising Addl Group of Surveillances in Which Justification Has Been Completed Following Receipt of 930623 Application ML20100J6831996-02-22022 February 1996 Application for Amend to License DPR-50,consisting of Change Request 254,revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Station Battery Cell Parameters Not Met ML20095J2261995-12-21021 December 1995 Revised TS Change Request 254,raising Low Voltage Action Level to 105 Volts DC ML20091L2831995-08-23023 August 1995 Rev to TS Change Request 223 to License DPR-50,incorporating Ref to 10CFR20.1302 ML20087F4881995-08-11011 August 1995 Rev to TS Change Request 252 to License DPR-50,removing TS Section 3.2 Re Makeup,Purification & Chemical Addition Sys Requirements.Certificate of Svc Encl ML20086R3001995-07-24024 July 1995 Rev to LAR 250 to License DPR-50,clarifying Section III, SE Justifying Change for source-range Nuclear Instrumentation SRs ML20085N1551995-06-22022 June 1995 TS Change Request 227 to License DPR-50,replacing Pages for App A.Certificate of Svc Encl ML20084N3461995-06-0101 June 1995 TS Change Request 223 to License DPR-50,deleting Remaining Portions of RETS & Relocating TSs Per Guidance in GL 89-01 & NUREG-1430 ML20084L4201995-06-0101 June 1995 TS Change Request 251 to DPR-50,revising TS 5.3.1.1 to Describe Use of Advanced Clad Assemblies ML20084B3261995-05-24024 May 1995 Application for Amend to License DPR-50.Amend Would Change Surveillance Test Requirements for source-range Nuclear Instrumentation ML20083N7771995-05-17017 May 1995 Application for Amend to License DPR-50,consisting of Change Request 252,removing Chemical Addition Sys Requirements from TS to COLR ML20079A7771994-12-23023 December 1994 Rev to TS Change Request 221 to License DPR-50,revising TS Page 3-32a ML20073F9241994-09-26026 September 1994 LAR 246 to License DPR-50,revising License Condition 2.C.9 Re Long Range Planning Program ML20069A6731994-05-20020 May 1994 TS Change Request 244 to License DPR-50,requesting Revised CR Trip Insertion Time Testing Acceptance Criterion for Remainder of Cycle 10 for 12 Specific CRs Which Initially Experienced Slow Rod Drop Times During Testing on 940317 ML20065M6681994-04-19019 April 1994 TS Change Request 237 to License DPR-50,deleting Audit Program Frequency Requirements from TS & Relocating Subj Requirements to Operational QA Plan ML20065K0381994-04-11011 April 1994 TS Change Request 238 to License DPR-50,relocating Detailed Insp Criteria,Methods & Frequencies of Containment Tendon Surveillance Program to FSAR ML20073C7701994-03-22022 March 1994 Application for Amend to License DPR-50 Re TS Change Request 242 Concerning Control Rod Trip Insertion Time Testing ML20064J3771994-03-11011 March 1994 Application for Amend to License DPR-50,revising TS Re Allowable Outage Time for Emergency Feedwater Pumps During Surveillance Activities 1999-08-20
[Table view] Category:TEXT-LICENSE APPLICATIONS & PERMITS
MONTHYEARML20211D1321999-08-20020 August 1999 Application for Amend to License DPR-50 to Revise TS Bases Re Degraded Undervoltage Relay Setpoint Calibration Frequency & Degraded Voltage Relay Tolerance Rev ML20210J0941999-07-29029 July 1999 Suppl TS Change Request 274 to License DPR-50,revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance Tss,As Result of NRC Generic Concerns Re CR Habitability Issues ML20209D6651999-07-0808 July 1999 Corrected TS Pages 3-21,4-5a,4-9,4-38 & 6-3 to Amends 211 & 212 to License DPR-50.Pages Failed to Reflect Previously Approved Changes Granted in Amends 203 & 204 in Case of Amend 211 & 207 in Case of Amend 212 ML20196K5901999-06-29029 June 1999 LAR 285 for License DPR-50,clarifying Authority in Encl License Pages to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-1 License to Amergen,Radioactive Matls Can Continue to Be Moved ML20209C0551999-06-29029 June 1999 LAR 77 for License DPR-73,clarifying Authority in Encl License Pages to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-1 License to Amergen,Radioactive Matl May Continue to Be Moved ML20195E3341999-06-0404 June 1999 TS Change Request 265 to License DPR-50,modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20209B5891999-05-26026 May 1999 LAR 281 to License DPR-50,revising TMI-1 UFSAR Chapters 5 & 14 to Permit Use of Conservative Deterministic Failure Margin Methodology for Seismic Analysis of Portions of Auxiliary Steam Line.Rept 240046-R-001 Encl.Rept Withheld ML20206R1111999-05-13013 May 1999 Application for Amend to License DPR-50,requesting That License Sections 2.a.(3) & 2.c.(7) & TS Section 3.1.1 Be Revised to Incorporate Administrative Updating & Changes to Bases Statement ML20205H0691999-04-0101 April 1999 TS Change Request 262 to License DPR-50,revising TS by Adding LCO Action Statements & Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20203B0421999-02-0202 February 1999 TS Change Request 274 to License DPR-50,expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems ML20196H9751998-12-0303 December 1998 LAR 279 to License DPR-50,reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196G4401998-12-0303 December 1998 LAR 278 to License DPR-50,requesting Consent to Transfer & Authorize Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License.Supporting Documentation,Encl ML20196G7671998-11-25025 November 1998 TS Change Request 277 to License DPR-50,changing Surveillance Specs for OTSG ISIs for TMI-1 Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only ML20195C6811998-11-12012 November 1998 Amend 52 to License DPR-73,relocating Audit Frequency Requirements from TMI-2 to TMI-2 Post Defueling Monitored Storage QA Plan & Extends Max Allowed Interval Between Certain Audits from 12 Months to 24 Months ML20154P8441998-10-19019 October 1998 Application for Amend to License DPR-50,providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage.Proprietary & non- Proprietary Calculation,Encl.Proprietary Info Withheld ML20154Q6201998-10-19019 October 1998 TS Change Request 248 to License DPR-50,adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18, Remote Shutdown Sys ML20154M3691998-10-15015 October 1998 Application for Amend to License Amend Request 276 to License DPR-50 Revising UFSAR Chapter 14 Accident Analysis Atmospheric Dispersion Factors.Ufsar Revised Pages Encl ML20249B2391998-06-11011 June 1998 TS Change Request 273 to License DPR-50,incorporating Alternate High Radiation Area Control Consistent W/Reg Guide 8.38 ML20217E5131998-03-23023 March 1998 Application for Amend to License DPR-50,revising TS Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 Efpy.Request Also Submitted in Response to NRC ML20217G0161997-10-0303 October 1997 Suppl to License Amend Request 268 to License DPR-50, Replacing Pages 4-80 & 4-81 Previously Submitted w/TMI-1 License Amend Request 268 on 970812 ML20211C2221997-09-19019 September 1997 Supplemental Application for Amend to License DPR-50 Re TS Change Request 266.Supplement Provides for Phased Review & Issuance of License Amends & Involves Administrative Correction to TS Decay Heat Removal Sys Leakage Rate ML20211C3211997-09-19019 September 1997 Supplemental Application for Amend to License DPR-50 Re License Amend Request 269 Submitted on 970814.Suppl Includes Addl Change to TS Section 3.1.4 & Proposed Page Rev Clarifying That UFSAR Analysis Is More Conservative ML20210J9951997-08-14014 August 1997 Revised License Amend Request 269 for License DPR-50, Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J3981997-08-12012 August 1997 Tech Spec Change Request 268 for License DPR-50,revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operations ML20198E7871997-07-30030 July 1997 TS Change Request 266 to License DPR-50,incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 ML20141E1311997-05-0808 May 1997 Application for Amend to License DPR-50,consisting of Change Request 264,incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2351997-04-21021 April 1997 Application for Amend to License DPR-50,requesting Rev to TS 3.3 Re Emergency Core Cooling,Reactor Bldg Emergency Cooling & Reactor Bldg Spray Systems ML20134M1111997-02-0707 February 1997 Application for Amend to License DPR-50,incorporating Certain Improvements from Revised STS for B&W Plants, NUREG-1430 & Addl Changes as Described in Encl TS ML20133D2651996-12-24024 December 1996 Application for Amend to License DPR-50,requesting Reinstatement of Auxiliary & Fuel Handling Bldg Ventilation Sys TS ML20132F3221996-12-16016 December 1996 Application for Amend to License DPR-50,consisting of Change Request 260,reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135C5201996-12-0202 December 1996 TS Change Request (Tscr) 75 to License DPR-73,relocating Audit Frequency Requirements to Pdms QA Plan ML20134D7761996-10-24024 October 1996 Amend 51 to License DPR-73,revising TS to Extend Surveillance Interval for Demonstrating Operability of Containment Airlock ML20128K1831996-10-0808 October 1996 Amend 50 to License DPR-73,revising TS to Incorporate Improvement from Administrative Controls Section of Revised STS for B&W Plants ML20117H0181996-08-29029 August 1996 Application for Amend to License DPR-50,consisting of TS Change Request 257,incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1901996-06-28028 June 1996 Application for Amend to License DPR-50,consisting of Change Request 259,allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8751996-05-24024 May 1996 Errata to Application for Amend to License DPR-50,consisting of TS Change Request 243,replacing Pages for App a ML20101R1361996-04-10010 April 1996 Application for Amend to License DPR-50,consisting of Change Request 243,revising Addl Group of Surveillances in Which Justification Has Been Completed Following Receipt of 930623 Application ML20100J6831996-02-22022 February 1996 Application for Amend to License DPR-50,consisting of Change Request 254,revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Station Battery Cell Parameters Not Met ML20095J2261995-12-21021 December 1995 Revised TS Change Request 254,raising Low Voltage Action Level to 105 Volts DC ML20091L2831995-08-23023 August 1995 Rev to TS Change Request 223 to License DPR-50,incorporating Ref to 10CFR20.1302 ML20087F4881995-08-11011 August 1995 Rev to TS Change Request 252 to License DPR-50,removing TS Section 3.2 Re Makeup,Purification & Chemical Addition Sys Requirements.Certificate of Svc Encl ML20086R3001995-07-24024 July 1995 Rev to LAR 250 to License DPR-50,clarifying Section III, SE Justifying Change for source-range Nuclear Instrumentation SRs ML20085N1551995-06-22022 June 1995 TS Change Request 227 to License DPR-50,replacing Pages for App A.Certificate of Svc Encl ML20084N3461995-06-0101 June 1995 TS Change Request 223 to License DPR-50,deleting Remaining Portions of RETS & Relocating TSs Per Guidance in GL 89-01 & NUREG-1430 ML20084L4201995-06-0101 June 1995 TS Change Request 251 to DPR-50,revising TS 5.3.1.1 to Describe Use of Advanced Clad Assemblies ML20084B3261995-05-24024 May 1995 Application for Amend to License DPR-50.Amend Would Change Surveillance Test Requirements for source-range Nuclear Instrumentation ML20083N7771995-05-17017 May 1995 Application for Amend to License DPR-50,consisting of Change Request 252,removing Chemical Addition Sys Requirements from TS to COLR ML20079A7771994-12-23023 December 1994 Rev to TS Change Request 221 to License DPR-50,revising TS Page 3-32a ML20073F9241994-09-26026 September 1994 LAR 246 to License DPR-50,revising License Condition 2.C.9 Re Long Range Planning Program ML20069A6731994-05-20020 May 1994 TS Change Request 244 to License DPR-50,requesting Revised CR Trip Insertion Time Testing Acceptance Criterion for Remainder of Cycle 10 for 12 Specific CRs Which Initially Experienced Slow Rod Drop Times During Testing on 940317 1999-08-20
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LICitT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY TilREE MILE ~ ISLAND NUCLEAR STATION UNIT I Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 88 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License tio. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.
METROPOLITAN EDISON COMPANY 07 ,
Sworn and subscribed to me this _ [/
day of _ h /97fp D
Notary Pt lic
[1 UCRCC J. TnorFER Nof:ry (m, nnen y,,g,, ca, 94 Comm!n :n Fypwn e,,,, y ym 790115007>
2
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY
'This is to certify' that a copy of Technical Specification Change Request No. 88 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, .
addressed as follows:
Mr. Weldon B. Arehart Mr. Harry B. Reese, Jr.
- Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. J. fl, Ceyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY I
By
_O/
Dated: 90
= - - -
, G. ' ,' '.
LThree Mile Island Nucicar Station, Unit I (TMI-1) e Operating License No. DPR-50
_ Docket No. 50-289 Technical Specification Change Request No. 88 The' licensee requests that the attached pages vi, 3-3, 3-4, and 3-5 replace like pages in the existing Technical Specifications, that the attached Figure 3.1-1 replace the current Technical Specification Figure 3.1-1, and that the current Figure 3.1-2 be deleted.
Reason for Change Request Metropolitan Edison Company (Met-Ed) was informed by the NSSS vendor, Babcock and Wilcox (B&W) that weld wire atypical of the submerged arc weld filler wire used .
by B6W in_the construction of reactor pressure vessels may have been used in the
- . construction of the TMI-1 Reactor Vessel. It is uncertain whether the atypical material has actually been used in the construction of the TMI-1 Reactor Vessel.
. Based on the assumption that the atypical material is present in the welds, B&W proposed the attached changes to Tech. Specs, which govern heatup and cooldown '
rates and pressure-temperature limits. . The commitment for this change request was made in LER 78-24/01T, dated August 21, 1978.
While the reactor coolant system heatup and cooldown limits were revised, it was determined that it was necessary to change the once-through steam generator (OTSC) pressure-temperature limit.
p Safety Evaluation Justifying Change There are two weld ' locations in the THI-1 Reactor Vessel that may contain the atypical material: upper shell to nozzle helt and lower.shell to the Dutchman.
Both are circumferential welds. The latter weld location is subject to low neutron fluence (nvt <1.0 x 1017 n/cm2 , E > 1 MeV) and the applied streuses are
- no greater than those applied in the beltline region.
A f racture mechanics evaluation ha- demonstrated that the structure integrity of the vessel' is uncompromised by th< aossible presence of the atypical material.
However, welds of the atypical material exhihit a higher unirradiated reference temperature, RTNDT. The f act that- RTNDT may have been underestimated would indicate that the ' current P-T limits would not have adequate margins of safety against non-ductile failure.
However, the applied pressure stress normal to the upper shell to nozzle belt weld is' approximately half that employed in calculating P-T limits, i.e., P-T limits were calculated assuming a defect oriented normal to the maximum stress direction.
-Thus, for circumferential welds,- the P-T limit curves have an added safety factor of 2. 'fith this safety factor and the safety factor of 2 applied on the stress intensi::y factor due to primary stresses (as required by' ASME Section III Appendix C),
the overall safety f actor is approximately four.-
s wt.
Aino, weldments of the atypical material have dif ferent silicon and nickel contents than the weldmenta using the atypical Mn 'fo-Ni weld wire. The copper content of the atypical weldments in about the name an the copper content of curveillance weldment s of Oconee Station, ifnits 2 and 3. Those nurveillance weldments exhibit a lower HTNDT than predicted by Reg. Guide 1.99 (Rev. 1).
The neutron induced embrittlement of the atypical weld meterial in expected to be equal to or less than the surveillance specimen weldmentn; equal because of the name copper content, possibly lower becaune of the lower nickel content. The effect of high at11 con is currently being evaluated by !!&W.
Furthermore, the current, and attached, P-T limits are applicable to only 5 EFPY.
Early in the life of the reactor, most of the P-T limita are controlled by the clonure head region during heatup and by the steam generator during con 19wn (see BAW 10046A, Rev. 1).
The revined operational heatup and cooldown limitn were calculated by assuming the atypical material van present in circumferential welda, and superimponing those points on exinting heatup and couldown curven. The operational limits include a safety factor of 2.0.
The operational 11mitu curve, with its safety f actor of 2.0, han been extrapolated to 2500 puin and will be used for hentup and cooldown for Innervice henk and liydrontatic tents as well.
The new limit for the OTSG was alno calculated uning proceduren dencribed in BAW-10046A , Rev. 1.
The temperature limit in baned on the highent RTNDT of the nteam generator (40"F) plun denign margin (60"F) plun a margin for inntrument error (12"F) for a total of 112"F. The preunure limit in based on 20% of the preoperational hydrostatic ten t prennure (1312 pntg) minun a margin for instrument error (24 pulg) for a total of 238 pnig.
The revised reactor coolant system heatup and cooldown limitu do not nubstantially differ from the current limit s, al though they are nomewhat more connervat ive.
Daned on the assumption that the atypical material may be present in nome weldn, the revised limits reflect the procedural requirements of BAW-10046A, Rev. 1. The revised reactor coolant system heatup and couldown 11mitu and OTSC prenuure-temperature limits will permit continued nafe operation of TMI-l in compliance with current regulations.
hicenne Amendmqnt Fee (10 CFl(17_0.22)
Because thin change involven a ningle innue, the licennee han determined that thin in a Clasu Ill Amendment and therefore the proper remittance in $4,000.
LIST OF FICURES Figure Title 2.1-1 Core Protection Safety Limit 2.1-2 Core Protection Safety Limits 2.1-3 Core Protection Safety Basis 2.3-1 Protection System Maximum Allowabic Set Points 2.3-2 Protection Syutem Maximum Allowable Set Points 3.1-1 Reactor Coolant System lleatup/Cooldown Limitations 3.1-2 ' Deleted 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter Il20.
3.5-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification Radial Flux '
Tilt Indication 3.5-2A- Rod Position Limits for 4 Pump Operation Applicable from 0 to 125+ EFPD; TMI-1, Cycle 4 3.5-2B Rod Position Limits for 4 Pump Operation from 125 + 5 EFPD to 265 i 15 EFPD; TMI-1, Cycle 4 3.5-2C Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period from 0 to 125 i 5 EFPD; TMT-1, Cycle 4 3.5-2D Rod Position Limits for 2 and 3 pump Operation from 125 1 5 EFPD to 265 i 15 EFPD; TMI-1, Cycle 4
.3.5-2E Power Imbalance Envelop for Operation from 0 to 125 i 5 EFFD, TMI-1, Cycle 4 vi L
3.1.2 PRESSURIZATION HEATUP AND COOLD0h'N LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.
Objective To assure that temperature end pressure changes in the reactor coolant system coolant system do not cause cyclic loads in excess of design for reactor components. * .
Specification ,
3.1.2.; For operations until five ef fective fuil power years, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of tha pressurizer) shall be limited in accordance with '
Figure 3.1-1 for normal operations and for inservice leak and hydrostatic testing.
Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1. Heatup and cooldown rates shall not exceed those shown in Figure 3.1-1.
3.1.2.2 The secondary side of the steam generator shall not be pre surized is above 238 psig if the temperature of the steam generator shell below 112 F.
exceed 1000F 3.1.2.3 The pressurizer heatup and cooldown rates shall not in any one hour. The spray shall not be used if the temperature ~
dif f erence between the pressurizer and the spray fluid is greater than 4300F.
3.1.2.4 Prior to exceeding five effective full power years of operation, Figure 3.1-1 shall be updated for the next service period [
The in accordance with 10 CFR 50, Appendix G, Section V.B.
highest predicted adjusted reference temperature of all the belt-line materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.5.
3-3 1
. 3.1.2.5 Tha upacecd proposed technical sprcifications referred to in 3.1.2.4 ch211 be cubmLtted for NRC review at least 90 days prior to the end of the cervice psriod. Appropriate additional NRC review time shall be allowed for proposed technical specification submitted in accordance with 10 CFR 50, Appendix G, Section V.C.
Bases All reactor coolant system components are designed to withstand the ef fects of cyclic loads due to system temperature and pressure changes. (1) These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR. The maximum temperature dependent heatup and cooldown rates are given below.
Max Heatup Rates Max Cooldown Ratas Temp 0F Max Rate Temp 0F Max Rate _
700 - 5500 500 F/hr Operating Range - 280 100'F/hr 280 - 150 500F/hr 150 - 2505/hr 1
The heatup and cooldown rate limits in this specification are not intended to limit instaneous rates of temperature change, but are intended to limit tempercture changes such that there exists no one hour interval in which a temperature change greater than the limit takes place.
The unirradiated reference temperature (RT NDT) for the surveillance region materials were determined in accordance with 10 CFR 50, Appendixes G and H.
For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using the methods described in BAW-10046A, Rev. 1.
As a result of f ast neutron irradiation in the beltline region of the core, The 1 there will be an increase in the RT NDT with accumulated neutron fluence. I adjusted reference temperatures have becc. calculated by adding the predicted .
radiation-induced ART NDT and the unirradiated RT NOT for each of the reactor coolant beltline materials.
The predicted RT NDT was calculated using the respective neutron fluence af ter five ef f ective f ull power years of operation. The upper limit of Figure 1 of Regulatory Cuide 1.99, Revision 1, was utilized for determining ART NDT. Fluence values are based on conservative analysis and extrapolation of actual measured neutron doses measured after the first fuel cycle to 5 ef fective full power years.
Based on the predicted RT NDT after five effective full power years of operation, the pressure-temperature limits of Figure 3.1-1 have been established in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAU-10046A, Rev. 1. The protection against nonductile failure is assured by maintaining the coolant pressure below the upper limits of these pressure-temperature limit curves.
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Th2 prassure limit lins on Figure 3.1-1 has been establishe@ considuring
'ths following:
- a. System pressure is measured in the loop with the pressurizar.
- b. Maximum differential pressure between the point of system pressure measurement and reactor vessel controlling regions for all operating pump combinations.
- c. No adjustment for instrument error, which is included in the procedure.
The spray temperature difference restriction, based on a stress analysis of spray line nozzle, is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of theof the steam gene steam generators. The highest estimated R + 60 F (100 F), the is 40 F. At metal temperatures lower than eved by limiting the protection against non-ductile failure isThe acsecondary coolant pressu highest estimated '
system hydrostatic test pressure of 1312 psig.of the steam RT The generato is gassumed to be equal to or greater than 0 the coolant temperature.
limits of 112 F0 and 238 psig include 12 F and 24 psig for possible in-strument error.
REFERENCES (1) FSAR, Section 4.1.2.4 0 (2) ASME Boiler and Pressure Code,Section III, N-415 (3) FSAR, Section 4.3.10.5 (4) BAW-1439, Analysis of Capsule TMI-IE from Metropolitan Edison Company, Three Mile Island Nuclear Station - Unit 1, Reactor Vessel Materials Surveillance Program.
,89 4
3-5
'r*.
l.. .
When Decay Heat Removal System is Assumed RTNDTe I Beltline 1/4T 238 -
operating without any RC pumps oper-Beltline 3/4T 177 -
ating, indicated Dit return temp. to Clocure Head Region 60 the Reactor Vessel shall be used. 60 Outlet Nozzle RC Pump Combinations Allowable:
Above 195 F All
!!ax Heatup Rate j 50 F/hr Below 195 F 1-A, 1-B; 0-A, 1-B; 1-A, 0-B Cooldown Rate:
Temp. C.D. Max Rate Operating Range-280 1000F/hr 280 -150 500 F/hr 150 - 250F/hr
. Point Press Temp.
A '105 75 410 125 G 2500 - B 550 210 A maximum step temperature C
D 550 275 change of 75 F is allowable 1500 300 F when removing all RC pumps
$! E
- F 2250 368 from operation with the G 2500 391 DHR system operating. The gj 2000 -
step temperature change is defined as th+. RC temp Dl gj (prior to stopping all RC
- pumps) ninus the DHR return temp (after stopping all RC h
- pumps). The above F tabulated ramp decrease is
% 1500 -
y allowable both before and after the step change.
S O
d a
Ej 1000 - '
0 n.
C D' 8 500 -
u B o
t; -
ef M A 100 200 300 400 500 REACTOR COOLANT TEMPERATURE, c T,F Reactor Coolant System Heatup/Cooldown Limitations (Applicabic to 5 EFPY)
Figure 3.1-1
t' 4 4 g
e h
FIGURE 3.1-2 DELETED
+
gs .e e G